This invention relates broadly to zirconium alloys, and more specifically to corrosion resistant and ductile zirconium alloys.
Improved zirconium and zirconium alloy nuclear fuel cladding tubes are disclosed in U.S. Pat. Nos. 4,200,492 and 4,372,817, incorporated herein by reference. A composite cladding tube described therein comprises a barrier layer of either high purity zirconium (such as crystal bar zirconium) or moderate purity zirconium (such as sponge zirconium) metallurgically bonded on the inside surface of a zirconium alloy tube. The composite cladding encloses the nuclear fuel material, leaving a gap between the fuel and the cladding. The barrier layer shields the alloy tube from the nuclear fuel material held in the cladding as well as shielding the alloy tube from fission products and gases. The barrier layer typically has a thickness equal to about 1 to about 30 percent of the thickness of the composite cladding. The barrier layer remains relatively soft during irradiation and minimizes localized strain inside the nuclear fuel elements, thus serving to protect the alloy tube from both stress corrosion cracking and liquid metal embrittlement. The alloy tube portion of the cladding is otherwise unchanged in design and function from previous practice for a nuclear reactor and is selected from conventional cladding material, such as zirconium alloys.
It is disclosed in U.S. Pat. Nos. 4,200,492 and 4,372,817 that the high and moderate purity zirconium metal forming the metal barrier in the composite cladding, even after prolonged irradiation, is able to maintain desirable structural properties such as yield strength and hardness at levels considerably lower than those of conventional zirconium alloys. In effect, the metal barrier does not harden as much as conventional zirconium alloys when subjected to irradiation, and this together with its initially low yield strength enables the metal barrier to deform plastically and relieve pellet-induced stresses in the fuel element during transients in neutron radiation. Fuel pellet induced stresses in the fuel element can be brought about, for example, by swelling of the pellets of nuclear fuel at reactor operating temperatures (300.degree. to 350.degree. C.) so that the pellet comes into contact with the cladding.
The nuclear fuel elements described in U.S. Pat. Nos. 4,200,492 and 4,372,817 provide a substantial improvement over elements which do not include internal zirconium barrier layers. However, the relatively pure zirconium barrier layers are subject to oxidation if the composite cladding is breached and water or steam enters the fuel rod during operation of the reactor. It would thus be desirable to improve the oxidation resistance in the barrier layer. It would be particularly desirable if such oxidation inhibition can be achieved without reducing the effectiveness of the zirconium barrier layer, particularly the ability of the barrier layer to deform plastically and relieve pellet-induced stresses in the fuel element during transients in radiation.
U.S. Pat. No. 4,894,203, incorporated herein by reference, discloses an improved nuclear fuel cladding having an alloy layer formed as a thin layer on the inner barrier layer, the alloy layer having less than one percent by weight of one or more impurities from the group consisting of iron, chromium, copper, nitrogen, and niobium. The alloy layer can be formed by any conventional method having control over the deposition depth of the impurity elements, such as ion implantation, ion plating, or chemical vapor deposition. The alloy layer provides improved oxidation resistance to the barrier layer without substantially affecting the desired plastic properties of the barrier layer.
U.S. Pat. No. 4,876,064, incorporated herein by reference, discloses corrosion resistant zirconium alloys comprised of 0.5 to 2.5 weight percent bismuth, or alternatively, 0.5 to 2.5 weight percent of a mixture of tin and bismuth, and 0.5 to 1.0 weight percent of a solute from the group consisting of niobium, molybdenum, tellurium, and mixtures thereof, and the balance zirconium. The alloys have a high resistance to both uniform and nodular corrosion as shown by laboratory high pressure steam testing that simulates the uniform and nodular corrosion that can be found on nuclear fuel rod cladding in the core of a nuclear reactor.
It is an object of this invention to provide a zirconium based alloy having an improved combination of corrosion resistance and ductility.
It is another object of this invention to provide a corrosion resistant zirconium alloy having ductility comparable to sponge zirconium.
It is another object of this invention to provide a corrosion resistant zirconium alloy having an improved combination of corrosion resistance and ductility suitable for barrier layers in nuclear fuel cladding.